ML20203D895

Proposed Tech Spec 6.3.1 Re Senior Reactor Operator Qualifications

ML20203D895
Person / Time
Site:	Catawba
Issue date:	07/09/1986
From:	DUKE POWER CO.
To:
Shared Package
ML20203D893	List: ML20203D889 ML20203D895 ML20203D900 ML20203D904
References
NUDOCS 8607210369
Download: ML20203D895 (3)

Text

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es 4 b ATTACHMENT 1 l

l ADNINISTRATIVE CONTROLS 6.2.3 CATAWA SAFETY REVIEW GROUP FUNCTIg 6.2.3.1 The Catawba Safety Review Group (CSRG) shall function to examine plant operating characteristics, NRC issuances, industry advisories, REPORTA8LE EVENTS, and other sources which may indicate areas for improving plant safety. The CSRG shall make detailed recommendations for revised pro-cedures, equipment modifications, or other means of improving plant safety to the Director, Nuclear Safety Review Board. j 1

COMPOSITION 6.2.3.2 The CSRG shall be composed of a chairman and at least four dedicated, full-time qualified individuals located onsite. ,

I RESPONSIBILITIES l 6.2.3.3 The CSRG shall be responsible for maintaining surveillance of plant activities to provide independent verification

• that these activities are performed correctly and that human errors are reduced as much as practical. j RECORDS .

6.2.3.4 Records of activities performed by the CSRG shall be prepared, maintained, and forwarued each calendar month to the Director, Nuclear Safety Review Board.

6.2.4 SHIFT TECHNICAL ADVISOR 6.2.4.1 The Shift Technical Advisor shall serve in an advisory capacity to l

the Shift Supervisor.

6.3 UNIT STAFF QUALIFICATIONS 6.3.1 Each member of'the unit staff shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable positions, except for the Radiation Protection Manager, who shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975. The licensed Operators and Senior Operators shall also meet or exceed the minimum qualifications of the supple-mental requirements specified in Sections A and C.of Enclosure 1 of the March 28, 1980 NRC letter to all licensees.**

6.4 TRAINING ,

6.4.1 A retraining and replacement training program for the unit staff shall .

be maintained under the direction of the Station Manager and shall meet or exceed the requirements and recommendations of Section 5.5 of ANSI N18.1-1971 and Appendix A of 10 CFR Part 55 and the supplemental requirements specified in Sections A and C of Enclosure 1 of the March 28, 1980 NRC letter to all >

licensees, and shall include familiarization with relevant industry operational experience identified by the CSRG.

• Not responsible for sign-off function.

• • Except that the experience and other considerations described in Duke Power Company's letters date.l August 28,1985 and July 8, 1986 are acceptable for the six and two applicants for SRO licenses identified therein, respectively.

CAT /MBA - UNITS 1 & 2 6-6 8607210369 860709 PDH ADOCK 05000413 p PDR

JUSTIFICATION AND ANALYSIS OF NO SIGNIFICANT HAZARDS CONSIDERATION The proposed change to Technical Specification 6.3.1 would permit an exception to the experience requirements for two candidates for senior reactor operator (SRO) licenses. The exception is from the requirements stated in Section A.1.a of Enclosure 1 to a letter from H. R. Denton to all licenses, dated March 28, 1980.

Section A.I.a of the Denton letter requires that an applicant for an SRO license have 4 years of experience as a control room operator (fossil or nuclear). This experience requirement is a prerequisite for taking the SRO examination. However, the primary requirement for an SRO license is successful ,

completion of the SRO examination.

Section A of the Denton letter allows exceptions to the experience requirements for SRO applicants at plants that are not yet licensed because there is no

! opportunity to obtain such experience on their plants. The proposed change to j Specification 6.3.1 is requested for a similar reason in that Catawba Unit 1,

which first received a fuel loading and precriticality testing license in July.

l 1984, a low power license in December,1984, and a full power license in )

January ,1985, has not been in operation long enough to provide an opportunity for reactor operators to have 4 years of control room operating experience.

l The two operators involved have similar experience and training as the six Catawba operators who were previously exempted from this requirement by  ;

Amendment No. 2 to license NPF-35 dated January 24, 1986.  :

) Technical Specification 6.3.1, as proposed, would acknowledge the Request for Waiver which was submitted to NRC/ Region II on July 8, 1986. As demonstrated by this letter the two SRO applicants have acceptable alternative experience.

i 10 CFR 50.92 states that a proposed amendment involves no significant hazards considerations if operation in accordance with the proposed amendment would not:

A l

(1) Involve a significant increase in the probability or consequences of  ;

an accident previously evaluated; or i (2) Create the possibility of a new or different kind of accident l previously evaluated; or (3) Involve a significant reduction in a margin of safety.

. The proposed amendment would not involve a significant increase in the probability or consequences of an accident previously evaluated because the two SRO candidates have been trained at Catawba, each has held a reactor operator license for more than one year and each would be required to pass the SRO examination.

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c The proposed amendment would not create the possibility of a new or different kind of accident from any accident previously evaluated because the SRO candidates are experienced, licensed operators and the amendment will not change the manner in which the plant is to be operated.

In addition, the proposed amendment would not involve a significant reduction in a margin of safety because the candidates have been trained at Catawba, they have been actively involved in startup testing and operation of both Catawba Units, and they will be required to pass the SRO examination.

Based upon this analysis, Duke Power Company concludes that the proposed caendment does not involve significant hazards considerations.

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